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A comparative study to investigate burnup in research reactor fuel using two independent experimental methods
Annals of Nuclear Energy 28 (2001) 1733±1744 www.elsevier.com/locate/anucene
A comparative study to investigate burnup in research reactor fuel using two independent experimental methods Masood Iqbal *, T. Mehmood, S.K. Ayazuddin, A. Salahuddin, S. Pervez Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, PO Nilore, Islamabad, Pakistan Received 9 December 2000; accepted 30 December 2000
Abstract Two independent experimental methods have been used for the comparative study of fuel burnup measurement in low enriched uranium, plate type research reactor. In the ®rst method a gamma ray activity ratio method was employed. An experimental setup was established for gamma ray scanning using prior calibrated high purity germanium detector. The computer software KORIGEN gave the theoretical support. In the second method reactivity dierence technique was used. At the same location in the same core con®guration the fresh and burned fuel element's reactivity worth was estimated. For theoretical estimated curve, group crosssections were generated using computer code WIMS-D/4, and three dimensional modeling was made by computer code CITATION. The measured burnup of dierent fuel elements using these methods were found to be in good agreement. # 2001 Elsevier Science Ltd. All rights reserved.
1. Introduction The burnup measurements for the nuclear reactor are important for the enhancement of safety, economics, and performance of a reactor. The main reason to perform such studies is due to the increased emphasis on reactor safety, increased cost of reactor fuels, and demand for higher neutron ¯uxes. The necessity for fuel and * Corresponding author. E-mail address: [email protected] (M. Iqbal). 0306-4549/01/$ - see front matter # 2001 Elsevier Science Ltd. All rights reserved. PII: S0306-4549(01)00013-5
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core management for more ecient fuel utilization by increasing burnup consistent with consideration of safety and performance became evident. To meet the demand for higher ¯uxes, using low enriched uranium (LEU) fuels, it became essential that detailed analyses of the new core be performed to optimize the available neutron ¯ux densities. Con®dence in core management scheme based on hand calculations and experience becomes much lower and less reliable and can be in some cases dangerous. Burnup determinations provide essential input to any systematic approach to the solution of the problem. In both reactivity eects and gamma ray spectrometry cases, methods depend upon the calculations, which support and supplement the measurements. The gamma ray spectrometry method, which measures speci®c ®ssion products is the direct way for measuring the burnup. The gamma ray spectrometry method needs a ®ne experimental arrangement for the fuel scanning. A well calibrated and high resolution detector is necessary for these measurements. The use of activity ratio technique in the gamma-ray spectroscopy method was ®rst suggested by Hick et al. (1970). The method has been fully investigated and discussed by Edder et al. (1973). Matsura (1975) also employed the gamma-ray spectroscopy method with a dierent way to determine the burnup for boiling water reactor. The reactivity eect method for the burnup measurements is now becoming popular for its simplicity and is less time consuming. Ravnik et al. (1992) described the reactivity method in detail and explained the drawbacks as well. They performed these measurements only for TRIGA fuel. Hammer et al. (1997) measured the burnup of the LEU fuel using the reactivity method for SAPHIR reactor. They estimated the accuracy of this method as 5%. Two non-destructive methods that is gamma ray spectrometry and reactivity eect have been employed for the burnup determination in Pakistan Research Reactor-I (PARR-1) LEU fuel. The PARR-I is a swimming pool type MTR, LEU fuel plate type research reactor. It became critical with LEU fuel in 1991. A set of three fuel elements was discharged in 1998. After the ®rst discharge of three fuel elements, the fuel burnup measurements were started. Initially reactivity eect measurements were performed. Then the gamma ray spectroscopy system was established and measurements were performed for the ®ssion product activity ratio. 2. Reactivity eect method For fuel burnup measurements using the reactivity method, a theoretical curve between reactivity worth and burnup of fuel element at the experimental location in the core is necessary. Based on this theoretical curve the experimentally measured worth of the fuel element was compared. Reactivity worth of a known burnup (fresh) fuel element is necessary for the reference. In the reactivity eect method the ®rst part is to measure the reactivity worth of the burned fuel elements at a particular location. If initially the reactor was made critical without any fuel element at the desired position in the core and is said to be in state K1 . Now after placing the burned fuel element at the particular location, the reactor becomes critical at a state
M. Iqbal et al. / Annals of Nuclear Energy 28 (2001) 1733±1744
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of K2 . The dierence of these two states K2 K1 will provide the worth of that fuel element. For reference to the fresh fuel element the same procedure can be applied. Based on the fresh fuel element worth, the calculated reference curve can be normalized. Using the normalized curve, the unknown burnup of the fuel element can be noted. For calculations, the energy group cross-sections were generated for the fuel element with dierent burnup using the WIMS-D/4 (Askew et al., 1966) computer code. WIMS-D/4 computer code is a general cell programme, which uses the transport theory to calculate the ¯ux as a function of neutron energy and position in the cell, group cross-sections and to solve the eigen value problem. Due to its generality, it is a very good tool for understanding the physics of the reactor lattice as it includes a large class of models and methods developed for reactor calculations as described by Iqbal (1995). Standard ¯ux weighted group collapsing was performed to obtain the condensed ®ve-energy-group constants for the required homogenized regions has already discussed by Iqbal (1995). For the whole reactor core, three dimensional calculations were done using an appropriate three dimensional model for the core in CITATION code (Fowler et al., 1971). The CITATION code is state of the art diusion theory based code, which treats a three dimensional criticality eigen value problem using ®nite dierence scheme. The neutron eigen value problem is solved by direct iterations to obtain the multiplication factor for the reactor system. Group constants from WIMS-D/4 for dierent regions were employed in CITATION as input to perform criticality search. Using these computer codes, a multiplication factor for the reference core as shown in Fig. 1 has been investigated by replacing the particular location (D-4) with dierent burnup fuel elements. The obtained theoretical curve for the system reactivity is depicted in Fig. 2. The system reactivity, due to the presence of the three burned and one fresh fuel element was measured experimentally at D-4 position in the core separately. For reactivity eect measurements, a control rod was calibrated. Initially the reactor was made critical without any fuel element at the D-4 position. The control rod position was noted. After that the fresh fuel element was loaded at the D-4 position and the control rod position was noted. Using ®rst and second reading, the system reactivity was calculated. The same procedure was repeated for all other fuel elements. Based on the theoretical curve the burnup of the burned fuel elements can be seen in Fig. 2. 3. Gamma-ray spectroscopy method In the gamma ray spectroscopy method activity ratio technique has been employed, in this method ®ssion product activity as a result of neutron ®ssion in the core is the main source. The reference ®ssion product that is burnup monitors employed are 95Zr, 134Cs, and 137Cs. The selection of the burnup monitors is very important and based on half life, ®ssion yield, and capture cross-section and other
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Fig. 1. Reference core con®guration employed in the reactivity eect technique.
Fig. 2. Theoretical calculated calibration curve between system reactivity and burnup for LEU fuel, also showing experimentally measured system reactivity for fuel elements.
M. Iqbal et al. / Annals of Nuclear Energy 28 (2001) 1733±1744
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factors as discussed below. 95Zr is produced through beta decay of mass-95 chain as follows
It has a comparatively short half life of 65.5 days. its activity reaches saturation after an irradiation of one year or so and, therefore, the amount present at the end of irradiation becomes a function of ¯ux alone i.e. 95
N t: :
1
where t> few half lives of 95 Zr. 134 Cs is a shielded nuclide and is produced only through neutron capture in 133Cs, a direct ®ssion product, according to the following scheme: 133 21h 133
I !
5:2d 133
Xe !
n; 134
Cs !
Cs
2:07y 134
!
Ba
It has a half life of 2.06 years, therefore, saturation activity is never achieved in normal operation of a reactor. Because of its production in two steps, the amount present at the end of irradiation is roughly a function of the square of the ¯uence, i.e. 134
N t: : :t2
2
137
Cs nuclide is produced by simple decay of the mass-l37 chain and has a long half life of 30.1 years. 137 24s 137
I !
4:2m 137
Xe !
Cs
30:1y 137
!
y 2:6m 137
Ba !
Ba
As it has negligible neutron capture cross-section, its amount present at the end of irradiation is a direct measure of the integrated number of ®ssions which have occurred in the fuel and hence corresponds to the burnup i.e. 137
N t: : :t
3
The ratios 95 Zr95/137Cs and 134 Cs/137Cs, used in this work and denoted by R1 and R2 respectively, have the following interesting features. From Eqs. (1) and (3) it can be inferred that the ratio R1 is proportional to the inverse of the irradiation time and, therefore, contains the information about the irradiation time. If the ¯ux is
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M. Iqbal et al. / Annals of Nuclear Energy 28 (2001) 1733±1744
known the irradiation time can be determined by `backward' calculations with the help of measured ratio R1 . From Eqs. (2) and (3) it is well clear that the ratio R2 is proportional to the ¯uence and hence contains the information about the burnup of the fuel. If the irradiation time is known the ¯ux can be obtained by theoretical `backward' calculations. It is possible to calculate these two ratios for any irradiation condition with considerable precision provided reliable nuclear data and relevant computer codes are available. These `forward' calculations would yield a number of irradiation conditions i.e. irradiation time and ¯ux, for a given measured ratio R1 or R2 . Since the measured ratios R1 and R2 for a fuel element represent a particular picture of ¯ux and irradiation time, a unique solution can easily be obtained by combining the dierent results yielded by the measured ratios. After calculating the ¯uence, burnup can be calculated using the usual formalism. A theoretical support is necessary for a unique value of integrated ¯ux for a particular fuel element based on the experimentally measured activity ratio. Based on this technique there is no need to consider the interim operation history and decay time because of the unique solution of two dierent ratios. The main purpose is to ®nd out the integral neutron ¯ux not neutron ¯ux or irradiation time. The accuracy of measurements of activity ratios depends only on the accuracy of the relative eciency of the detector for dierent energies. These eciency ratios are easily obtained using a standard point radiation source commonly available in the laboratory. Other factors in¯uencing the activity are gamma-ray attenuation in material between the source and detector. 3.1. Experimental setup Experimental setup as shown in Fig. 3 was installed. A HPGe detector was used for the gamma-ray spectroscopy. The proper energy calibration and eciency calibration were done prior to the measurements. Special arrangement was made for the fuel element movement. Sucient shielding was arranged for the better results of the detector. Each fuel element was placed in front of the detector separately. For axial scanning there were six steps for each fuel element (10 cm). The data of time 500 s was stored for each step. The background was also recorded. The typical spectrum of one measurement for the fuel element is shown in Fig. 4. After the completion of the experiment the data were analyzed. 4. Data analysis Firstly the counts under the gamma-ray energy peaks of interest were obtained from the stored data and shown in Tables 1±3. The corrections were made for these data sets. These corrections include (a) gamma-ray intensities, (b) cooling time, and (c) gamma-ray attenuation in the fuel element, pool water and gamma cell window. The material based data were taken from engineering compendium by Jaeger et al. (1970). Then activity ratios were further corrected using the energy eciency ratio. The corrected factors due to attenuation and energy eciency are shown in Table 4.
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Fig. 3. Experimental setup for the fuel element gamma-ray spectroscopy.
The corrected activity ratio used for the burnup measurement is shown in Table 5. The mean value of the activity ratio for each fuel element was used for the burnup in that fuel element. The computer code KORIGEN (Fischer et al., 1983) which is an improved Karlsruhe version of ORNL isotope generation and depletion code ORIGEN was used for the theoretical calculations for the base of these measurements. The KORIGEN nuclear data library suitable for PARR-l fuel and three spectral indices has already been used in another report by Bukhtyar et al. (1996). The calculations were performed for dierent neutron thermal ¯ux and dierent time of irradiation. The dierent activity ratios were calculated against dierent ¯uence for particular thermal ¯ux and are shown in Fig. 5. From these graphs ¯uence and thermal ¯ux were taken against the experimentally measured activity ratio. The graph against
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Fig.4. Typical energy spectrum obtained from the LEU spent fuel after cooling of 493 days. Table 1 Measured count rate at dierent positions of fuel element No. S-69 at required gamma-ray energy peaks Distance from top (cm)
137
5 15 25 35 45 55
206.02 351.15 492.55 543.05 573.02 351.05
a
Cs 662 (keV) (0.42)a (0.33) (0.29) (0.28) (0.28) (0.35)
95
134
Zr
Cs
724 (keV)
757 (keV)
605 (keV)
796 (keV)
7.26 11.62 14.42 14.79 15.65 10.19
8.25 12.99 16.64 16.27 17.69 11.17
19.13 53.99 100.36 121.11 117.79 44.68
16.30 49.34 91.82 112.10 108.95 44.39
(4.24) (3.67) (3.80) (4.18) (4.20) (5.24)
(3.88) (3.25) (3.17) (3.63) (3.56) (4.67)
(3.24) (1.03) (1.14) (1.03) (1.10) (2.21)
(2.02) (1.10) (0.79) (0.73) (0.76) (1.37)
Standard % errors are in parentheses.
thermal ¯ux and ¯uence was drawn for both activity ratios as shown in Fig. 6. The cross-section of both curves provides the actual value of neutron ¯uence, which was used, for the burnup calculations in the formula given below. M M0 1 e f t where
M. Iqbal et al. / Annals of Nuclear Energy 28 (2001) 1733±1744
M M0 f t
= = = =
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U-235 mass burnt initial mass of U-235 spectrum averaged ®ssion cross-section neutron ¯uence.
Table 2 Measured count rate at dierent positions of fuel element No. S-73 at required gamma-ray energy peaks Distance from top (cm) 5 15 25 35 45 55 a
137
95
Cs 662 (keV)
323.00 567.68 597.55 491.21 425.44 240.21
134
Zr
a
(0.37) (0.27) (0.27) (0.28) (0.29) (0.38)
Cs
724 (keV)
757 (keV)
605 (keV)
796 (keV)
11.00 17.66 19.01 17.28 17.78 11.08
11.74 20.07 20.41 19.46 20.49 12.75
49.73 130.36 148.04 115.71 70.36 22.77
49.40 119.94 139.52 106.01 60.19 18.05
(4.87) (3.67) (3.44) (3.28) (2.79) (3.23)
(4.48) (3.14) (3.16) (2.84) (2.42) (2.85)
(1.92) (0.96) (0.89) (1.00) (1.40) (2.95)
(1.24) (0.71) (0.64) (0.72) (1.05) (2.09)
Standard % errors are in parentheses.
Table 3 Measured count rate at dierent positions of fuel element No. S-69 at required gamma-ray energy peaks Distance from top (cm)
5 15 25 35 45 55 a
137
Cs 662 (keV)
299.80 471.21 554.70 535.30 455.53 333.49
a
(0.33) (0.27) (0.27) (0.29) (0.32) (0.36)
95
134
Zr
Cs
724 (keV)
757 (keV)
605 (keV)
796 (keV)
12.51 16.34 16.24 14.99 11.93 9.90
14.67 19.86 19.53 17.07 12.45 10.14
36.17 87.59 128.47 124.12 81.03 38.46
35.25 85.16 120.28 114.54 74.86 36.90
(3.37) (3.180 (3.77) (4.23) (5.03) (5.23)
(2.50) (2.60) (3.01) (3.61) (4.87) (5.09)
(1.95) (1.16) (0.97) (1.01) (1.41) (2.39)
(1.15) (0.72) (0.68) (0.76) (1.04) (1.65)
Standard % errors are in parentheses.
Table 4 Corrected values of activity ratio for three fuel elements Distance from top (cm)
S-69 Zr/137Cs
10 20 30 40 50 60 Average
S-73 CS/137CS
Zr/137Cs
S-77
95
134
95
134
CS/137CS
95
Zr/137Cs
9.92 9.25 8.30 7.55 7.67 8.04 8.46
0.12 0.20 0.26 0.29 0.26 0.17 0.21
9.32 8.76 8.73 9.86 11.85 13.06 10.26
0.20 0.29 0.32 0.30 0.21 0.11 0.24
11.91 10.07 8.47 7.89 7.10 7.98 8.90
134
CS/137CS
0.16 0.24 0.30 0.30 0.23 0.15 0.23
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Fig. 5. Activity ratio vs neutron ¯uence for dierent thermal neutron ¯ux.
M. Iqbal et al. / Annals of Nuclear Energy 28 (2001) 1733±1744
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Table 5 Gamma-ray attenuation factor and energy eciency factor used for the data correction Energy ratio
Attenuation factor due to material
Energy eciency factor
605/662 796/662 724/662 757/662
1.2072 0.6485 0.81696 0.73455
1.1089 0.7556 0.8904 0.8361
Fig. 6. Solution curve for neutron ¯ux and ¯uence to get unique solution for ¯uence.
Table 6 Experimentally measured burnup for three fuel elements using two independent methods Fuel element number
Gamma spectroscopy method (%)
Reactivity eect method (%)
Ratio
S-69 S-73 S-77
34.14 36.98 35.16
31.60 36.89 32.23
1.08 1.002 1.09
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5. Conclusions Two independent experimental methods (reactivity eect method and gamma-ray spectroscopy method) were employed for fuel burnup measurements in plate type LEU fuel of PARR-1. In reactivity eect method theoretical support was added using WIMSD/4 and CITATION neutronic computer codes. in gamma-ray spectroscopy method fuel depletion KORIGEN computer code was used. Looking at Tables 1±3, the count rate of the ®ssion products indicate almost a cosine shape while in Table 4 the corrected activity ratio of 95 Zr/137Cs did not follow this pattern. This can be discussed as the dierence in the burnup of these ®ssion products. The results of burnup in each fuel element from both methods are shown in Table 6. The result shows a good agreement between these two methods. Gamma ray spectroscopy method needs a lot of experimental set up exercise and is complicated as compared to the reactivity method. Also for gamma-ray spectroscopy method sucient decay time is necessary for experimental measurements. For saturation in 95 Zr minimum irradiation of 400 days is also necessary to attain saturation activity. For early and easy way to measure the fuel burnup in LEU plate type fuel elements, reactivity eect method can be used easily. Acknowledgements The authors would like to thank the operation sta of PARR-1 for their continuous help during experimentation. We are also indebted for the assistance of Mrs Sabiha Bukhtyar during theoretical calculations. References Askew, J.R. et al., 1966. A general description of lattice code WIMS. J. Brit. Nucl. Energy Soc. 5, 564. Bukhtyar, S., et al., 1996. Burnup dependent isotopic concentrations and decay parameters of PARR-1 LEU silicide fuel. PINSTECH-154. Edder, O.J. et al., 1973. SM-170/12, Symp. Proc. Nuclear Data in Science and Technology, Paris, Vol 1, IAEA, Vienna, p. 233. Fischer, U, et al., 1983. KORIGEN-Verbesserte konsistente berechnung des nuklearen inventars abgebrannter DWR-brennstoe auf der basis von Zell-Abbrand-Verfahren mit KORIGEN, KFK-3014. Fowler, T.B. et al., 1971. Nuclear reactor core analysis code: CITATION, ORNL-TM-2496, Rev. 2. Oak Ridge National Laboratory. Hammer, J. et al., 1997. Burnup determination of LEU fuel at SAPHIR Reactor. ANL/RERTR/TM-18. Hick, H., et al., 1970. SM-133/5, Symp. Proc. Progress in Safeguards Technology, Karlsruhe, p. 533. Iqbal, M., 1995. Comparison of experimental and calculated neutron spectra for Pakistan Research Reactor-1. Nucl. Sci. Journal 32, 5. Jaeger, G. et al., 1970. Engineering Compendium on Radiation Shielding, Vol. II. Springer-Verlag, New York. Masura, T., 1975. Non-destructive gamma-ray spectrscopy on spent fuels of a boiling water reactors. J. Nucl. Sci. Technol 12, 24. Ravnik, M. et al., 1992. Determination of burn-up of TRIGA fuel elements by calculation and reactivity experiments. Kerntechnik 57, 5.
A comparative study to investigate burnup in research reactor fuel using two independent experimental methods Masood Iqbal *, T. Mehmood, S.K. Ayazuddin, A. Salahuddin, S. Pervez Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, PO Nilore, Islamabad, Pakistan Received 9 December 2000; accepted 30 December 2000
Abstract Two independent experimental methods have been used for the comparative study of fuel burnup measurement in low enriched uranium, plate type research reactor. In the ®rst method a gamma ray activity ratio method was employed. An experimental setup was established for gamma ray scanning using prior calibrated high purity germanium detector. The computer software KORIGEN gave the theoretical support. In the second method reactivity dierence technique was used. At the same location in the same core con®guration the fresh and burned fuel element's reactivity worth was estimated. For theoretical estimated curve, group crosssections were generated using computer code WIMS-D/4, and three dimensional modeling was made by computer code CITATION. The measured burnup of dierent fuel elements using these methods were found to be in good agreement. # 2001 Elsevier Science Ltd. All rights reserved.
1. Introduction The burnup measurements for the nuclear reactor are important for the enhancement of safety, economics, and performance of a reactor. The main reason to perform such studies is due to the increased emphasis on reactor safety, increased cost of reactor fuels, and demand for higher neutron ¯uxes. The necessity for fuel and * Corresponding author. E-mail address: [email protected] (M. Iqbal). 0306-4549/01/$ - see front matter # 2001 Elsevier Science Ltd. All rights reserved. PII: S0306-4549(01)00013-5
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core management for more ecient fuel utilization by increasing burnup consistent with consideration of safety and performance became evident. To meet the demand for higher ¯uxes, using low enriched uranium (LEU) fuels, it became essential that detailed analyses of the new core be performed to optimize the available neutron ¯ux densities. Con®dence in core management scheme based on hand calculations and experience becomes much lower and less reliable and can be in some cases dangerous. Burnup determinations provide essential input to any systematic approach to the solution of the problem. In both reactivity eects and gamma ray spectrometry cases, methods depend upon the calculations, which support and supplement the measurements. The gamma ray spectrometry method, which measures speci®c ®ssion products is the direct way for measuring the burnup. The gamma ray spectrometry method needs a ®ne experimental arrangement for the fuel scanning. A well calibrated and high resolution detector is necessary for these measurements. The use of activity ratio technique in the gamma-ray spectroscopy method was ®rst suggested by Hick et al. (1970). The method has been fully investigated and discussed by Edder et al. (1973). Matsura (1975) also employed the gamma-ray spectroscopy method with a dierent way to determine the burnup for boiling water reactor. The reactivity eect method for the burnup measurements is now becoming popular for its simplicity and is less time consuming. Ravnik et al. (1992) described the reactivity method in detail and explained the drawbacks as well. They performed these measurements only for TRIGA fuel. Hammer et al. (1997) measured the burnup of the LEU fuel using the reactivity method for SAPHIR reactor. They estimated the accuracy of this method as 5%. Two non-destructive methods that is gamma ray spectrometry and reactivity eect have been employed for the burnup determination in Pakistan Research Reactor-I (PARR-1) LEU fuel. The PARR-I is a swimming pool type MTR, LEU fuel plate type research reactor. It became critical with LEU fuel in 1991. A set of three fuel elements was discharged in 1998. After the ®rst discharge of three fuel elements, the fuel burnup measurements were started. Initially reactivity eect measurements were performed. Then the gamma ray spectroscopy system was established and measurements were performed for the ®ssion product activity ratio. 2. Reactivity eect method For fuel burnup measurements using the reactivity method, a theoretical curve between reactivity worth and burnup of fuel element at the experimental location in the core is necessary. Based on this theoretical curve the experimentally measured worth of the fuel element was compared. Reactivity worth of a known burnup (fresh) fuel element is necessary for the reference. In the reactivity eect method the ®rst part is to measure the reactivity worth of the burned fuel elements at a particular location. If initially the reactor was made critical without any fuel element at the desired position in the core and is said to be in state K1 . Now after placing the burned fuel element at the particular location, the reactor becomes critical at a state
M. Iqbal et al. / Annals of Nuclear Energy 28 (2001) 1733±1744
1735
of K2 . The dierence of these two states K2 K1 will provide the worth of that fuel element. For reference to the fresh fuel element the same procedure can be applied. Based on the fresh fuel element worth, the calculated reference curve can be normalized. Using the normalized curve, the unknown burnup of the fuel element can be noted. For calculations, the energy group cross-sections were generated for the fuel element with dierent burnup using the WIMS-D/4 (Askew et al., 1966) computer code. WIMS-D/4 computer code is a general cell programme, which uses the transport theory to calculate the ¯ux as a function of neutron energy and position in the cell, group cross-sections and to solve the eigen value problem. Due to its generality, it is a very good tool for understanding the physics of the reactor lattice as it includes a large class of models and methods developed for reactor calculations as described by Iqbal (1995). Standard ¯ux weighted group collapsing was performed to obtain the condensed ®ve-energy-group constants for the required homogenized regions has already discussed by Iqbal (1995). For the whole reactor core, three dimensional calculations were done using an appropriate three dimensional model for the core in CITATION code (Fowler et al., 1971). The CITATION code is state of the art diusion theory based code, which treats a three dimensional criticality eigen value problem using ®nite dierence scheme. The neutron eigen value problem is solved by direct iterations to obtain the multiplication factor for the reactor system. Group constants from WIMS-D/4 for dierent regions were employed in CITATION as input to perform criticality search. Using these computer codes, a multiplication factor for the reference core as shown in Fig. 1 has been investigated by replacing the particular location (D-4) with dierent burnup fuel elements. The obtained theoretical curve for the system reactivity is depicted in Fig. 2. The system reactivity, due to the presence of the three burned and one fresh fuel element was measured experimentally at D-4 position in the core separately. For reactivity eect measurements, a control rod was calibrated. Initially the reactor was made critical without any fuel element at the D-4 position. The control rod position was noted. After that the fresh fuel element was loaded at the D-4 position and the control rod position was noted. Using ®rst and second reading, the system reactivity was calculated. The same procedure was repeated for all other fuel elements. Based on the theoretical curve the burnup of the burned fuel elements can be seen in Fig. 2. 3. Gamma-ray spectroscopy method In the gamma ray spectroscopy method activity ratio technique has been employed, in this method ®ssion product activity as a result of neutron ®ssion in the core is the main source. The reference ®ssion product that is burnup monitors employed are 95Zr, 134Cs, and 137Cs. The selection of the burnup monitors is very important and based on half life, ®ssion yield, and capture cross-section and other
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M. Iqbal et al. / Annals of Nuclear Energy 28 (2001) 1733±1744
Fig. 1. Reference core con®guration employed in the reactivity eect technique.
Fig. 2. Theoretical calculated calibration curve between system reactivity and burnup for LEU fuel, also showing experimentally measured system reactivity for fuel elements.
M. Iqbal et al. / Annals of Nuclear Energy 28 (2001) 1733±1744
1737
factors as discussed below. 95Zr is produced through beta decay of mass-95 chain as follows
It has a comparatively short half life of 65.5 days. its activity reaches saturation after an irradiation of one year or so and, therefore, the amount present at the end of irradiation becomes a function of ¯ux alone i.e. 95
N t: :
1
where t> few half lives of 95 Zr. 134 Cs is a shielded nuclide and is produced only through neutron capture in 133Cs, a direct ®ssion product, according to the following scheme: 133 21h 133
I !
5:2d 133
Xe !
n; 134
Cs !
Cs
2:07y 134
!
Ba
It has a half life of 2.06 years, therefore, saturation activity is never achieved in normal operation of a reactor. Because of its production in two steps, the amount present at the end of irradiation is roughly a function of the square of the ¯uence, i.e. 134
N t: : :t2
2
137
Cs nuclide is produced by simple decay of the mass-l37 chain and has a long half life of 30.1 years. 137 24s 137
I !
4:2m 137
Xe !
Cs
30:1y 137
!
y 2:6m 137
Ba !
Ba
As it has negligible neutron capture cross-section, its amount present at the end of irradiation is a direct measure of the integrated number of ®ssions which have occurred in the fuel and hence corresponds to the burnup i.e. 137
N t: : :t
3
The ratios 95 Zr95/137Cs and 134 Cs/137Cs, used in this work and denoted by R1 and R2 respectively, have the following interesting features. From Eqs. (1) and (3) it can be inferred that the ratio R1 is proportional to the inverse of the irradiation time and, therefore, contains the information about the irradiation time. If the ¯ux is
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known the irradiation time can be determined by `backward' calculations with the help of measured ratio R1 . From Eqs. (2) and (3) it is well clear that the ratio R2 is proportional to the ¯uence and hence contains the information about the burnup of the fuel. If the irradiation time is known the ¯ux can be obtained by theoretical `backward' calculations. It is possible to calculate these two ratios for any irradiation condition with considerable precision provided reliable nuclear data and relevant computer codes are available. These `forward' calculations would yield a number of irradiation conditions i.e. irradiation time and ¯ux, for a given measured ratio R1 or R2 . Since the measured ratios R1 and R2 for a fuel element represent a particular picture of ¯ux and irradiation time, a unique solution can easily be obtained by combining the dierent results yielded by the measured ratios. After calculating the ¯uence, burnup can be calculated using the usual formalism. A theoretical support is necessary for a unique value of integrated ¯ux for a particular fuel element based on the experimentally measured activity ratio. Based on this technique there is no need to consider the interim operation history and decay time because of the unique solution of two dierent ratios. The main purpose is to ®nd out the integral neutron ¯ux not neutron ¯ux or irradiation time. The accuracy of measurements of activity ratios depends only on the accuracy of the relative eciency of the detector for dierent energies. These eciency ratios are easily obtained using a standard point radiation source commonly available in the laboratory. Other factors in¯uencing the activity are gamma-ray attenuation in material between the source and detector. 3.1. Experimental setup Experimental setup as shown in Fig. 3 was installed. A HPGe detector was used for the gamma-ray spectroscopy. The proper energy calibration and eciency calibration were done prior to the measurements. Special arrangement was made for the fuel element movement. Sucient shielding was arranged for the better results of the detector. Each fuel element was placed in front of the detector separately. For axial scanning there were six steps for each fuel element (10 cm). The data of time 500 s was stored for each step. The background was also recorded. The typical spectrum of one measurement for the fuel element is shown in Fig. 4. After the completion of the experiment the data were analyzed. 4. Data analysis Firstly the counts under the gamma-ray energy peaks of interest were obtained from the stored data and shown in Tables 1±3. The corrections were made for these data sets. These corrections include (a) gamma-ray intensities, (b) cooling time, and (c) gamma-ray attenuation in the fuel element, pool water and gamma cell window. The material based data were taken from engineering compendium by Jaeger et al. (1970). Then activity ratios were further corrected using the energy eciency ratio. The corrected factors due to attenuation and energy eciency are shown in Table 4.
M. Iqbal et al. / Annals of Nuclear Energy 28 (2001) 1733±1744
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Fig. 3. Experimental setup for the fuel element gamma-ray spectroscopy.
The corrected activity ratio used for the burnup measurement is shown in Table 5. The mean value of the activity ratio for each fuel element was used for the burnup in that fuel element. The computer code KORIGEN (Fischer et al., 1983) which is an improved Karlsruhe version of ORNL isotope generation and depletion code ORIGEN was used for the theoretical calculations for the base of these measurements. The KORIGEN nuclear data library suitable for PARR-l fuel and three spectral indices has already been used in another report by Bukhtyar et al. (1996). The calculations were performed for dierent neutron thermal ¯ux and dierent time of irradiation. The dierent activity ratios were calculated against dierent ¯uence for particular thermal ¯ux and are shown in Fig. 5. From these graphs ¯uence and thermal ¯ux were taken against the experimentally measured activity ratio. The graph against
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Fig.4. Typical energy spectrum obtained from the LEU spent fuel after cooling of 493 days. Table 1 Measured count rate at dierent positions of fuel element No. S-69 at required gamma-ray energy peaks Distance from top (cm)
137
5 15 25 35 45 55
206.02 351.15 492.55 543.05 573.02 351.05
a
Cs 662 (keV) (0.42)a (0.33) (0.29) (0.28) (0.28) (0.35)
95
134
Zr
Cs
724 (keV)
757 (keV)
605 (keV)
796 (keV)
7.26 11.62 14.42 14.79 15.65 10.19
8.25 12.99 16.64 16.27 17.69 11.17
19.13 53.99 100.36 121.11 117.79 44.68
16.30 49.34 91.82 112.10 108.95 44.39
(4.24) (3.67) (3.80) (4.18) (4.20) (5.24)
(3.88) (3.25) (3.17) (3.63) (3.56) (4.67)
(3.24) (1.03) (1.14) (1.03) (1.10) (2.21)
(2.02) (1.10) (0.79) (0.73) (0.76) (1.37)
Standard % errors are in parentheses.
thermal ¯ux and ¯uence was drawn for both activity ratios as shown in Fig. 6. The cross-section of both curves provides the actual value of neutron ¯uence, which was used, for the burnup calculations in the formula given below. M M0 1 e f t where
M. Iqbal et al. / Annals of Nuclear Energy 28 (2001) 1733±1744
M M0 f t
= = = =
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U-235 mass burnt initial mass of U-235 spectrum averaged ®ssion cross-section neutron ¯uence.
Table 2 Measured count rate at dierent positions of fuel element No. S-73 at required gamma-ray energy peaks Distance from top (cm) 5 15 25 35 45 55 a
137
95
Cs 662 (keV)
323.00 567.68 597.55 491.21 425.44 240.21
134
Zr
a
(0.37) (0.27) (0.27) (0.28) (0.29) (0.38)
Cs
724 (keV)
757 (keV)
605 (keV)
796 (keV)
11.00 17.66 19.01 17.28 17.78 11.08
11.74 20.07 20.41 19.46 20.49 12.75
49.73 130.36 148.04 115.71 70.36 22.77
49.40 119.94 139.52 106.01 60.19 18.05
(4.87) (3.67) (3.44) (3.28) (2.79) (3.23)
(4.48) (3.14) (3.16) (2.84) (2.42) (2.85)
(1.92) (0.96) (0.89) (1.00) (1.40) (2.95)
(1.24) (0.71) (0.64) (0.72) (1.05) (2.09)
Standard % errors are in parentheses.
Table 3 Measured count rate at dierent positions of fuel element No. S-69 at required gamma-ray energy peaks Distance from top (cm)
5 15 25 35 45 55 a
137
Cs 662 (keV)
299.80 471.21 554.70 535.30 455.53 333.49
a
(0.33) (0.27) (0.27) (0.29) (0.32) (0.36)
95
134
Zr
Cs
724 (keV)
757 (keV)
605 (keV)
796 (keV)
12.51 16.34 16.24 14.99 11.93 9.90
14.67 19.86 19.53 17.07 12.45 10.14
36.17 87.59 128.47 124.12 81.03 38.46
35.25 85.16 120.28 114.54 74.86 36.90
(3.37) (3.180 (3.77) (4.23) (5.03) (5.23)
(2.50) (2.60) (3.01) (3.61) (4.87) (5.09)
(1.95) (1.16) (0.97) (1.01) (1.41) (2.39)
(1.15) (0.72) (0.68) (0.76) (1.04) (1.65)
Standard % errors are in parentheses.
Table 4 Corrected values of activity ratio for three fuel elements Distance from top (cm)
S-69 Zr/137Cs
10 20 30 40 50 60 Average
S-73 CS/137CS
Zr/137Cs
S-77
95
134
95
134
CS/137CS
95
Zr/137Cs
9.92 9.25 8.30 7.55 7.67 8.04 8.46
0.12 0.20 0.26 0.29 0.26 0.17 0.21
9.32 8.76 8.73 9.86 11.85 13.06 10.26
0.20 0.29 0.32 0.30 0.21 0.11 0.24
11.91 10.07 8.47 7.89 7.10 7.98 8.90
134
CS/137CS
0.16 0.24 0.30 0.30 0.23 0.15 0.23
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Fig. 5. Activity ratio vs neutron ¯uence for dierent thermal neutron ¯ux.
M. Iqbal et al. / Annals of Nuclear Energy 28 (2001) 1733±1744
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Table 5 Gamma-ray attenuation factor and energy eciency factor used for the data correction Energy ratio
Attenuation factor due to material
Energy eciency factor
605/662 796/662 724/662 757/662
1.2072 0.6485 0.81696 0.73455
1.1089 0.7556 0.8904 0.8361
Fig. 6. Solution curve for neutron ¯ux and ¯uence to get unique solution for ¯uence.
Table 6 Experimentally measured burnup for three fuel elements using two independent methods Fuel element number
Gamma spectroscopy method (%)
Reactivity eect method (%)
Ratio
S-69 S-73 S-77
34.14 36.98 35.16
31.60 36.89 32.23
1.08 1.002 1.09
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5. Conclusions Two independent experimental methods (reactivity eect method and gamma-ray spectroscopy method) were employed for fuel burnup measurements in plate type LEU fuel of PARR-1. In reactivity eect method theoretical support was added using WIMSD/4 and CITATION neutronic computer codes. in gamma-ray spectroscopy method fuel depletion KORIGEN computer code was used. Looking at Tables 1±3, the count rate of the ®ssion products indicate almost a cosine shape while in Table 4 the corrected activity ratio of 95 Zr/137Cs did not follow this pattern. This can be discussed as the dierence in the burnup of these ®ssion products. The results of burnup in each fuel element from both methods are shown in Table 6. The result shows a good agreement between these two methods. Gamma ray spectroscopy method needs a lot of experimental set up exercise and is complicated as compared to the reactivity method. Also for gamma-ray spectroscopy method sucient decay time is necessary for experimental measurements. For saturation in 95 Zr minimum irradiation of 400 days is also necessary to attain saturation activity. For early and easy way to measure the fuel burnup in LEU plate type fuel elements, reactivity eect method can be used easily. Acknowledgements The authors would like to thank the operation sta of PARR-1 for their continuous help during experimentation. We are also indebted for the assistance of Mrs Sabiha Bukhtyar during theoretical calculations. References Askew, J.R. et al., 1966. A general description of lattice code WIMS. J. Brit. Nucl. Energy Soc. 5, 564. Bukhtyar, S., et al., 1996. Burnup dependent isotopic concentrations and decay parameters of PARR-1 LEU silicide fuel. PINSTECH-154. Edder, O.J. et al., 1973. SM-170/12, Symp. Proc. Nuclear Data in Science and Technology, Paris, Vol 1, IAEA, Vienna, p. 233. Fischer, U, et al., 1983. KORIGEN-Verbesserte konsistente berechnung des nuklearen inventars abgebrannter DWR-brennstoe auf der basis von Zell-Abbrand-Verfahren mit KORIGEN, KFK-3014. Fowler, T.B. et al., 1971. Nuclear reactor core analysis code: CITATION, ORNL-TM-2496, Rev. 2. Oak Ridge National Laboratory. Hammer, J. et al., 1997. Burnup determination of LEU fuel at SAPHIR Reactor. ANL/RERTR/TM-18. Hick, H., et al., 1970. SM-133/5, Symp. Proc. Progress in Safeguards Technology, Karlsruhe, p. 533. Iqbal, M., 1995. Comparison of experimental and calculated neutron spectra for Pakistan Research Reactor-1. Nucl. Sci. Journal 32, 5. Jaeger, G. et al., 1970. Engineering Compendium on Radiation Shielding, Vol. II. Springer-Verlag, New York. Masura, T., 1975. Non-destructive gamma-ray spectrscopy on spent fuels of a boiling water reactors. J. Nucl. Sci. Technol 12, 24. Ravnik, M. et al., 1992. Determination of burn-up of TRIGA fuel elements by calculation and reactivity experiments. Kerntechnik 57, 5.